• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.280-291
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• 2018

In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010-2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1464-1470
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• 2022

This paper presents a high performance burnable absorber named as CIMBA (Cylindrically Inserted and Mechanically Separated Burnable Absorber) for the soluble-boron-free SMR. The CIMBA is the cylindrical gadolinia inserted into the annular fuel pellets. Although the CIMBA utilizes the spatial self-shielding effect of the fuel material, a large reactivity upswing occurs when the gadolinia is depleted. To minimize the reactivity swing of the CIMBA-loaded FA, two approaches were investigated. One is controlling the spatial self-shielding effect of the CIMBA as burnup proceeds by a multi-layered structure of the CIMBA (ML-CIMBA) and the other is the mixed-loading of two different types of CIMBA (MIX-CIMBA). Both approaches show promising performances to minimize the reactivity swing, where the MIX-CIMBA is more preferable due to its simpler fabrication process. In conclusion, the MIX-CIMBA is expected to accelerate the commercialization of the CIMBA and can be used to achieve an optimal soluble-boron-free SMR core design.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.105-109
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• 2000

종래의 핵연료소결체가 혼합산화물 혹은 이산화우라늄중 한가지 핵연료만으로 구성한 것과 달리 내부를 저농축 이산화우라늄 핵연료로 채우고 그 외부를 링형태의 혼합산화물 핵연료로 둘러 싼 이중구조를 특징으로 한다. 이러한 형태의 핵연료소결체는 중심영역의 핵분열반응률 줄임으로써 핵분열 기체생성, 핵연료봉 중심온도와 평균온도를 낮추어 준다. 이는 핵분열 기체방출을 낮추어 혼합산화물 핵연료봉 성능을 향상시키고 방출 연소도를 증가시키는 효과가 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1075-1076
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2008.05a
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• pp.173-174
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• 2008

• Proceedings of the Korean Institute of Resources Recycling Conference
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• 2004.05a
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• pp.138-142
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• 2004

• Proceedings of the Korean Institute of Resources Recycling Conference
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• 2003.10a
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• pp.206-210
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.321-322
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• 2005

• Nuclear Engineering and Technology
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• v.12 no.4
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• pp.274-285
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• 1980

The TDCORE and the RELOAD-II codes are adopted in an attempt to establish a simplified computational system for the in-core fuel management decisions of a PWR. The TDCORE is being used to simulate the power and burnup behavior of the Gori unit 1 PWR during the fuel cycle 1 through the cycle 5. The validity of the TDCORE code is also presented by comparing the TDCORE prediction with the in-core measurements. The RELOAD-II code is used to determine the optimum fuel loading pattern which is one of the most important decisions of the Gori unit 1 reactor. The utility and applicability of two codes for the fuel management analyses are described.